Neutronic reactor



July. 15, 1958 v R. F. CHRISTY 2,843,543

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R. F. CHRISTY NEUTRONIC REACTOR July 15, 1958 12 sheets-sneu e Y Filed Oct. 19, 1945 FIEE.

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NEUToNIc REAcToR Filed oct. 19, 1945 12 sheets-sheet fr FIE..

July 15, 1958 R. F. CHRISTYv v 2,843,543

` NEuTRoNIc REACTOR Filed Oct. 19. 1945- 12 Sheets-Sheet 8 n July 15,v 1958.

Filed oct( 19. 1945 R. F. CHRISTY NEUTRONIC REAC'IOR 12 Sheets-Sheet 9 Julyv 15, 1958 l R. F. CHRlsTY 2,843,543

NEUTRoNIc REACTOR Filed Oct. 19, l945 v l2 Sheets-Sheet lO July 15, 1958 I l R. F. CHRISTY 2,843,543

' NEUTRONIC REACTOR Film1v oct. 19, 1945 12 sheets-sheet 11 s o :l I l l l l I l l l l 0^ if) N w 4 E I Q 5 I E l 3 I lo I 2' E 5 I (Bm-404 I v U Beo(p=3)(4o5) IQZ I I I l I I I I I I I I I I 0 2o 4o 6o 8o RADIUS oF SPHERE (CM) w/ N12-5555; iNvmoR. 1 Q gf Robert E Chr/sry R. F. CHRISTY 2,843,543

NEuTRoNIc REAcToR July l5, 1958 12 Sheets-Sheet l2 Filed 00T.. 19, 1945 o Q o 1 E l N o I M .SQHOND mso .0325/u x N l o /owm r Nsowmowo S .mi Ow nog nog

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INVENTOR. Robert E Christy v five centimeters,

jUnited States Pate-Ht Y l 2,843,543 NEUTRoNIc nEACToR Robert F. christy, sama Fe, N; Mex., assigner tothe United States of America as represented by the United p States Atomic Energy Commission n Application ctober 19, 1945, Serial VNo. 623,363 2S Claims. (Cl. 204-193.2)

i atomic numbers is approximately equal totheatomic number ofthe original nucleus.Y However, a mass defect exists and a consideration of the process kin terms of the conservation of energy reveals that a substantial amountV f of energy is released during the ssion process. Furtherymore,V on the average, more than one fast neutron is emitted for every neutron absorbed to initiate Vfission.

' Itis therefore clear that, if the fast neutrons produced byV vfission can be vmade to cause new iissions in such propor- .tiony that the overall neutron generation exceeds the overall losses in and from the system, the ,chain reaction can be divergent to a desired rate of neutron generation. n As a 7 consequence, the energy released during the fission process is'available in the form of heat and/or radiation forexl v tended periods of time; that is, during the continuanceof energy for thelchain reaction. The employment of that useful purposes forms the basis of this invention.

The secondary neutrons produced by the lissioning of a issionable isotope nucleus have a high average energy.

y More specifically the mean lenergy in the fissionr neutronV spectrum is inthe neighborhood of from 0.5 to 3 million 'l electron volts (M. E. V.), and the mean free path of such neutrons in a substantially solid vmass of a iissionable isotopeis comparatively short, for example, yof the order of the result being that the mean time be- Y f tween fissions in such an arrangement will be ofthe order Y passing them through a of a'hundredth of a microsvecond.V While a fast neutron chain reaction can 'be maintained in such anV arrangement if a sufficient quantity of such a ssionable isotope'or material is brought together in lfavorable geometry, i. e., aVV

Qquantity in excess of the critical-mass Value, it has-'been determined that for the purposes ofthe present invention, the employment of thermal neutrons to produce ssions 'permits of a number of advantages.

The requirement that the neutrons employed in a controlled neutron chain reaction of the type contemplated by this invention' be slowed to near thermal energies by slowing medium called a moderator in whichy they are slowed by atomic collisionsgarises f Aout of the following considerations:

l As will 'be vshown 1n more detail later, 'the cross section I for ssion (i. e. the probability that fission will occur under Y neutron bombardment) increasesy as the energy of the primary or incident neutronis reduced.` In fact, the cross section is approximately inversely proportional to the neutron energy'. vA-s a concomitant factor, the variousmoderators have diterent efficiencies as neutron slowing media ,as well as different absorption cross sections for neutrons.

' By the'p'roper choice of f in which theneutrons are Y energies and which absorb very a'moderating material, that is Ione quickly slowed to thermal few neutrons, it is possible ICC Y c Patented July 15,1958

to take advantage of the increased cross section for fission c' of the ssionable isotope and thereby reduce by as muchV tion. If on the other hand, the fission neutrons can fbe slowed down to thermal energies vand a'chainvreaction initiated, since the mean time between fissions in sucha reaction will be great, suicient control over the reaction can be readily maintained and the desired rate lof neutron generation fixed at any desired level.

Isotopes that have been determined to be Vappropriate for slow neutron chain reaction include, for example,

' isotopes of uranium (element 92) havingthe atomic weights 233 and 235 and isotope of plutonium (element 94) having Vthe atomic weightV 239. These ssionable' isotopes have no substantial threshold for the energy of the incident neutron, hence vfission may be initiated by f a slowor thermal neutron, i. e., a neutron whose energy is approximately that of thermal agitation.

It might be noted also that, for a substantial part of the energy spectrum, the cross-section' for fissionwfor these isotopes is almost inversely proportional or primary neutron energy, that is,.the cross-section approximately follows the i law. Variousmixtures of these isotopes in elementaL or compound form and mixtures with other elements or isotopes can be used when following the teachings of the present invention,-as will be explained hereinafter.

' Fermi and Szilard in U. S. Patent No. 2,708,656 issued May 17, V1955, haveY disclosed .methods andvmeans Yfor establishing slow neutron chain reactions rwhich continue in aself-sustaining manner at predetermined levels of neutron density. VThe system there disclosed provided for the Aemployment -of uranium in itsA normal polyisotopic state,- that is, uranium .238 admixedwith approximately 0.7 percent of uranium 235, as the ssionable. material.

' Other component elements which form what is now known as aneutronic reactor system include:

tor, such as graphite in which the issionable lmaterial isV ,dispersed in a geometrical pattern designed to reduce neutron losses. A Y Y Y f 'j 1 (2) Heat removal means for example, channels l-in heat Vexchange relationship with the reacting mass and through which a suitable'coolant is circulated in order to stabilize temperatures in the system. f I l n (3) Anlouter casing lwhich serves toreectneutrons back into the systemr andk therebynreduce lthe Yquantity (i. e. the Vcritical mass) of fissionable mixture,necessary` to sustain the reaction. This outer casingrfis-sometimes termed a tamper. Y

(4) Means for charging the-reactive-,elements into`the zone inwhich the reactiontakes place` and for removal therefrom of the products of the reaction.` 1 i (5) A protective shield ispsometimres provided around the reactor to minimize the escape o f biologically harmful radiations. Such shields may comprise, vvfor example, Y Vbisnzruthor lead which havebeen found elective in stop'- ping Vgamma radiation, hydrogenous materials such as paran for absorbing-neutrons and/ or aY massiveiouter concrete casing. if` g (6)'A monitoring .system toV determine the Areaction u conditions at all times. L Y f 1 `(7) 'Controldevices gen'erallycomprisingV neutron absorbing s materials insertable into the 'reactive mass to to the Vincident; Y

maintainanayerage state of neutron production and absorption balance at a predetermined level.

(8) A safety device comprising :a quantity of neutron absorbing materialrwhich may be used to stop the reaction in case., of' emergency by being automatically inserted into neutron-absorbing relationship with the reacting mass.

Ingconsider'ingQthe requirements for an operating neutronic.,reactongtheratio of secondary neutrons produced bythe ffissions; tothe original number of primary neutrons of thetype required to initiate the tis-sions in a chain 'reacting'systemofg'infnite size using speciiic materials is called-,the reproduction or multiplication factor of the System.V AThe-factor is a dimensionless constant and is denotedbythesymbol k. If k is made sufficiently greater thanfunity-togcreate a net gain in neutrons over all interior ;losses, .andfthe system is of proper size so that this gain is not entirely lost by leakage from the exterior surfacemothe system, then a self-sustaining chain reactingsyst-em canbe built, to generate neutrons and to produce p owerin-:the form of heat by nuclear fission. .Importantlossesof neutrons within the reacting mass havebeeii found to be by absorption in contaminating impur-itieswhich are present with the iissionable mixture (e. g;.polyisotopiczuranium) or by absorption in uranium 238.,.w1`thout :producing iission but instead, leading to the formation of plutonium 239 as will be explained later. The .absorptionby the contaminating materials varies, but .thee'ifectlio'n the k factor may be readily determined hylthelemployment of formulae disclosed in the above mentioned; application. The effect of numerous elements has been correlated in this way with the composition of iissionable material and moderator or neutron slowing material. That is, for example, more normal polyisotopic uranium can b e added to a particular system to overcome 'thewabsorptionl effects of impurities in the system.

yUfran'iurn-238has an especially strong absorbing power for neutrons'iivhili have been slowed to moderate energies. The energy levels at which this absorption is strongest''are 'known as resonance energies, and the neutron-'capt-uiefor absorption by uranium 238 nuclei at these ,energies isi-therefore known as the uranium resonance 'captureor'absorption Such absorption is to be distinguishedfromfbsorption in impurities as discussed above. l ffhe'setwo-neutron loss factors are most important in the-determination-of whether fa -self-sustaining chain reaction*canfbeniaintained. Together with the loss of neutron" by @leakage out of the system, the above mentioned losses govern the size of the reactor.V Thus reactorsconstrfucte'df according to prior art principles have beencomparatively large, massive installations requiring L extremely large-quantities of the various elements and/ or materials described above.

AIt should-falso fbe noted that the eiiiciency of a neutron slowing 'niate'rialor moderator,'depends upon .the scatteringi cross-section -bf the material vand its atomic weight. Thus, for example, hydrogen has a high scattering crosssection-and;'allowL atomic weight and is :an extremely desirable.neutroni-slowing agent because of the small numberfofatorniccllisions necessary to slow a neutron to therr'na'l=energes-2lv When present in the'form of'water however, the absorption cross-section fis comparatively high-and4the5k`factors for uranium and water are very close-to unityandjthe advantages ofthe use of the hydrogen are largely-lost.

Ithasbee'n pointed out above that control means are prvide'd'infiectors for stabilizing the neutron density at -predeterminedI levels. ASuch controls have normally been in theforrn'of neutron absorbing materials inserted rdirectly intdthereacting mass, thuseectively taking the neutrons'ldirectlytout of the reaction. Such controls are subjectedto a,V great deal of fast neutron Aas well as'therl malnentrn bombardment andV means foi-,coolingthem have, been founduseful if not completely essential.

Itwillthu b'efseen thatvthis invention has as an object the"'provisioh"of'a method and means for establishing a self-sustaining slow neutron chain reaction in a compact unit suitable for general use.

It is a further object of the present invention to provide a means and method of so co-relating the essential physical requirements of a fission chain reaction that practical and easily controllable neutronic reactors can be built.

It is a still further object of the present invention to provide means for producing neutrons and radiations for transmutation purposes.

Another object of the present invention is to provide a novel method and means for controlling a self-sustain ing slow neutron chain reaction.

Another object of the present invention is to provide a reactor system in which the multiplication factor is independent of most of the neutron losses generally encountered in such a system.

Other objects and advantages will become apparent from the disscussion in this speciiication and from the detailed description of illustrative embodiments which are given by way of example and should not be interpreted to be limitations of the broad pinciples underlying the invention.

The above mentioned objects and advantages are attained by employing a composition of a iissionable isotope and moderator in iiuid form, for example, one in which the fissionable isotope is suspended or preferably dissolved in a liquid moderator such as water or heavy water (i. e. deuterium oxide, D20). In such an arrangement the amount of the iissionable isotope present, to a large extent, governs the reaction and eliminates the problems attendant upon complex impurity removal techniques and the like. ln other words, by the use of the methods and principles herein disclosed, the neutron absorption effect caused by (a) the presence of impurities, (b) isotopes which absorb neutrons without resulting in fission, (c) absorption in the moderator, (d) absorption by iission products and like effects, can be readily overcome by the novel expedient of increasing the concentration of the specific fissionable isotope present in the system. Thus, higher neutron losses can be tolerated than is the case when natural polyisotopic uranium is used, but losses still can be overcome to the end that a self-sustaining chain reaction canvbe maintained. As a consequence, the size of the reactor is no longer a critical factor, the new criterion being the concentration of the tissionable isotope.

It has been noted that among the materials which can be employed in the practice of the present invention arc the uranium isotopes of mass 233 and 235 and the plutonium isotope 239, all of which have no substantial threshold for the energy of the incident neutron. These isotopes can be obtained in highly concentrated form by isotopic separation procedures or chemical methods (depending on the isotope or element) and brief mention is made here of such methods as background for this invention and to emphasize further benefits derived from following the novel methods and using the apparatus herein described.

The iissionable isotope uranium 235 may be obtained in several ways. Isotope separation devices such as a mass spectroseparator, similar in operation to a mass spectrograph but with larger ion currents, have been found satisfactory. Another method of separating the uranium 235 isotope from the naturally occurring isotopic mixture is by gas diffusion methods employing uranium hexafluoride gas and diffusion barriers. In both methods the separation is not completed in a single stage, but rather proceeds step-wise, or in cascade fashion, with the accepted portion of each step being further separated and the rejected portion being recycled. It will thus bc secr: that the fissionable isotope is observed to occur in greater abundance or concentration, with each advancing step in -the process,

Uranium 233 may be formed by subjecting a quantity forth.

(standpoint of the present invention.

90232 n 9023s i, ,Y

' 23.5 min.

27.4 days 91233 922334.

If desired, the uranium 233 can be separated from the thorium parent by chemical methods but as will bel seen from the discussion herein, this separation is not necessary if the concentration of. the Vuranium isotope is sufficiently high according to the standards hereinafter set Plutonium 239 is formed principally by irradiation of uranium 238 with neutrons. LAs a production method, one way of subjecting large quantities of uranium to a high neutron flux is the employment of a reactor such as is disclosed in the above-mentioned patent of Fermi and 'Szilard Therreaction leading to the formation of plu- Since the plutonium yis. formed in the original uranium slugs dispersed in the graphite reactor, chemical extraction and/orprecipitation processes may be used to obtain the isotope 239 in a substantially pure state, but here again complete separation is not necessary from the It is generally preferred in the practice of the present invention to employ a water soluble salt containing the fissionable isotope in the desired isotopic concentration orin a substantiallypure isotopic state. For example, uranyl salts of high water solubility such as uranyl nitrate,

' uranyl sulphate or uranyl'uorideQplutonium salts Ysuch LA aS4 plutonyl sulphate (PuO2SO4), plutanyl nitrate (PuO2(NO3)-2), plutonous Ynitrate (Pu(NO3)4), etc. may

V be dissolvedin water and used in the neutron chain reaction herein contemplated.

YIt will be apparent to one skilled in the art that employing `a composition of a materialenriched inla iissionable isoltope With a water moderator, and following the practices and standards hereinafter set forth, it is possible to vary the neutron gain (that is, yvary the muln'plication factor k) by increasing the concentrationV of the iissionable `isotope in agiven volume. 'It has 'been determined that the neutron losses due to the presence of an absorbing isotope such as uranium 238 can be made relatively unimportant without eliminating the uranium238 from the system. Thus, if anV isotopic mixture of uranium 235 and uranium 238-is employed, if the concentration of uranium 235 is sutflciently above that of natural uranium, the lossespdueto absorption of neutrons by the uranium 238 becomes'negligible and can be neglected in the design -of a'reactor, particularly one using a moderator of high neutron absorption properties such'as water even though the amount of uranium 238 is high. It' has been determined further, that lwhere the concentration of uranium 235 is above about one percent and preferably above tive percent Yby Weight of the uranium present, a great reduction in the amount of uranium in optimum geometry necessary to establish ay l chain reaction (i. e. the critical mass of uranium required) canV be effected, and 'comple'teor substantially complete removal of absorbing isotopes or impurities is unnecessary. For example, if a water moderator is used and the 'fissionable mixture is the normal isotopic uranium mixture reaction can-be established at all. vBy Way lof omparif son. if the enrihmentis 180%'-(the Vurramuttl ,235v .contant about two percent of the uranium present), only about 1.7 tons of uraniumare requiredunder the sameicondig tions of operation.` Even more striking is the determination that when the uranium composition contains fifteen but the use ofsuch reflectorsdoes not affect the general principles here noted. Y n Y Y The critical mass'values lfor a reactorrof substantially spherical geometry, as well as the critical dimensions and concentration of the ssionable Yisotopeand'the interdependence of these criteria .for iissionable isotopes such as have been mentioned, may be calculated as follows: The neutron distribution in a reactor as a function of the radius of the reactor is the solution of the diffusion equation: A 'y f (kP,-`1) l A1t|L2 71,-() Y (l) where n is the neutron density, An, where A is the Loplocian operator, is defined by the relation:

521|, 52uY Sn n saa t? for a system with cartesian coordinates x, y and z, Pt isY gies before leaking out of the reacting mass, k is the re-` production factor for an innitemedium, and L is the thermal diffusion length of the neutrons in the dispersion of the ssionable isotope in moderator, (for a spherical homogeneous reactor whererf is the radius of the reactor) is written sin Krv then K2L2=kPt(K.) -l

where L0 is the thermal diffusion length of the neutrons in the pure moder-ator and UNAM) and ast(M) are respectively the thermal neutron absorption and thermal neu# tron scattering cross-sections of the moderator. The'second term in the denominator fis a usually negligible correction to the total cross-section.

meter of the reacting solution. Also VeX 1 -l-X where Ve is the effective number of neutrons per thermal fission of the iissionable isotope and vincludesthejadditional neutrons formedby fast lission andis'further 'dened by the -relation f K 5.5.,M &f F P hence the tissioning) cross-section vof the ssionable isotopes respectively, ra't(E). is theabsorption cross-section of the iissionable isotope Vforether'mal neutrons andVis Y' Y If the solution It is assumed Vthat theV presence of the iission'able .isotope does not appreciably 'change the number of hydrogen nuclei per cubic centiwhere GMM) and UMF) are the'fast neutron scattering l cross-section of the moderator and the absorption (and Y the actual number of neutrons produced per fission. The term fast fission includes the range where the fission cross-section is essentially constant, i. e., from 10,000 e. v. up to fission energies. Or stated another way, it was assumed that the fast fissioncross section of about 1 Barn (10*24 cm?) remained constant down to an energy E, expressed in electron volts and defined by UMF) l #40E The region of energies greater than E was taken for the fast group. The number of collisions necessary to slow a neutron to thermal energies is then where E is the mean natural logarithmic energy decrement per collision -in the moderating medium, P1(K) is the average probability of escaping leakage for these energies. Then Expanding the denominator on the left, one gets a quadratic equation for X.

When Fermis concept of neutron age applies in the slowing vdown precedure, so that the distribution of nascent thermal neutrons from a point source of fast neutrons can Vbe written T2 e *n in which r is a distance from the source, then Pt(K)=e-K21 and P1 K =fKZf1 r is the neutron age which is 1/6 of the mean square displacement of a neutron from place of birth to the point at which the neutron reaches the energy for which the computations are to be made. -r1 is the appropriate age of the fast neutrons making fast fission and is the range of the neutron for the first few collisions. In water, the distribution of energetic neutrons from a fission source is awr After the first few collisions, the distribution spreads in an approximately Gaussian manner with an age r from this lower energy to thermal energies. This consideration leads tanrlKl and for a slab the thickness fails This permits calculation of the critical mass, mass/cm., and mass/cm.2 of plutonium 239, for example, respectively for a sphere, cylinder, and slab, as a function of the density of plutonium 239, or as a function of the dimensions.

Except for the region of large density, the critical mass of uranium 235 or uranium 233 is greater than that of plutonium 239 `by the factor UadPum) UMUZSS) i. e., by 1.7 or l for the same dimensions of the mixture.

Since the function of the moderating medium, i. e. water, heavy water (D20) or the other low atomic number element having a low capture cross-section, is to slow the fission neutrons, the critical size will be of the order of the slowing down distance. The minimum concentration is such that only one of the 2.13 effective neutrons per absorption in a uranium 235 nucleus and 1.98 effective neutrons per absorption in a plutonium 239 nucleus is absorbed by the chain reactive fissionable isotope compound, the thermal neutron absorption by the fissionable material will then be about equal to that by the moderator; the optimum concentration (minimum critical mass in a sphere) will be about three times this minimum.

The control of a neutronic reactor is an important factor, since if the reaction is permitted to occur at an unduly rapid rate the reaction will take place with explosive violence. Control of a neutronic reaction may be eected by variation of one or more of the above losses or by variation in the concentration of fissionable isotope. For example, the reactor may be controlled by introducing into or withdrawing from the reaction zone high neutron absorbers such as cadmium or boron usually in the form of control rods.

In accordance with the present invention it has been found that a neutronic reactor may `be effectively controlled by variation of the leakage from the reacting composition. Thus a neutronic reactor has been con structed which is below critical size, i. e., the size of the reactor is so small that leakage of neutrons from the reacting composition without a reflector is too great and this loss alo-ne prevents establishment of a neutronic` self-sustaining reaction. But when this reactor is provided with a reflector which reflects enough neutrons back into the reaction zone to reduce the leakage loss, a point can be reached such that a self-sustaining neutron chain reaction can be established. The reflector is also provided with means to vary the amount of neutrons so reflected. For example such a means may comprise one orimore neutron absorbing-'controlrods whichmayv be removablyr inserted in the V'reflector to absorbfneutrons therein. As another means to4 accomplish this'purpose, Yfer example, a portion of the reflector Vmaybe blocked off .by neutron absorbers if desiredor the amount of rey llector or its Ydepth may be varied.rj At all *events thereactor may be controlled by lcontrol of, ltheleakage factor t'eri'stic time vfor neutron generation basedfupon theper'- cent of enrichment of iissionable isotope employed-Vin the composition with the moderator,.the type of moderator, vthe reflector used and the like.; Thi'sfeharaeteristic tiniemay be used as a base to which may be'relatedy the determination Aof whether the neutrons emitted `in the fission process are prompt or delayed t In the ssion of'yuranium 23S about one percent maybe .termedv delayedneutrons, although the percentage variesforthe different isotopes. These*r delayed fast *neutrons may appear at any time up toseveral minutes after the tission has occurred. In uranium 235 for example, half of theseV neutrons "are emitted .within six vseconds .and- 0.9 within 'Y y45 seconds. YThe mean time of delayed emission lis' about 5 seconds. The neutron reproduction cycle is completed by 99 percent of the neutrons in about 0.00003'second in a lfluid type reactor Asystem employing a watervmodera- .tor such as forms the basis of the present invention, 'a1- i though the dependence of this value on the moderator chosen should be noted. But if the reactor isoperating with a reproduction ratio near unity, the extra one per; if cent may make all the diierence between :an increase'or a decrease in the activity of the reactor. The fact that the last neutron in thercycle is held back,.as it were, im-

` parts a slowness of response by the reactor-systemtothe Vchanges in the control meansthat would not be present if the fission neutrons were Vall emitted instantaneously.

For cases in which the reproduction ratio (Rrr) differs safety. s Y

-from unity by appreciably less than one percent, rthe .rise

of neutron density, or more speciically the ,valueNnto which the number of neutrons has risen from an original Y -value No, after a lapse of time of Yt seconds during and before which the pile has operatedtat a iixedvalue ofRV (N0 being the number of neutrons at the beginning of t,

' Y .has settled down to aA steady 'f vand'applicable for low'values i. e., after disappearance'of transient eie-cts vdue to any' preceding change in R), is given'byN- i-Noewwhere vw -R1 l v a-(R-l) T I, In this `formula .a is the fraction of the neutrons that are delayed, e. g., in the case of the uranium 235 isotope oc=0.0067, T is the mean delay'time for'the delayed n'eu of the system. The above formula is'V only approximate of R 'becauseiit ;usesY an average delay time. 3"

As an example,ysuppose as a result of' moving! l,

shows how the reduction ofthe rate of the delayed `neutron effect isparticularly significant in the statedlower rangev o'f'R values.' Strictly speaking, the given equation holds only for the steady state, i. e., where R has been held constant for vsome time; an additional transient term must be Vincluded to obtain an accurate representation of the neutron `density during the first few seconds afterfa sudden change of R.

If R were made Vexactly 1.0067, a more detailed theory shows that the neutron density would be more than tripled each second. However, if the reproduction. ratio Y,R is several percent greater than unity, so that the one percent delayed neutrons are unimportant compared with R-1, the density increases at a much more rapid rate as given approximatelyV by "(R-0.0067.)tl where l is 0.00003 second, `the normal' time to complete'a cycle in a 'reactor such as is described hereinafter. Thus if R were to be made '1.04, the neutron density would increase in 0.03

second by a factor of approximately 101'1 over its original level. .HowevergifR were 1.02 and 1.03, the factor by which the neutrondensity would bejmultiplied eachv second, would be 1100 and 700,000 respectively. VItis thus app'arentthat too' high a reproduction ratio ina practical system leads to the necessity of inserting what may be considered as an excessive amount of ,controlling absorber's to reduce the eiective'reproduction ratio to unity.

l as 4with a higher ratio, only at a slower rate, thelower reproduction ratios which `exceed uinty by not substantially morethan about 0.01, or an amount equal to the percentage of the neutrons' formed which are delayed neutrons are preferred in practice inthe interest Yof Y -nle-appueafion of the principlesV ser ,forth just above to neutronic reactors utilizing high concentrations ofr iis# 'sionable isotopes, will'be -more fully understood by ret'-V erence to the drawings wherein a preferred embodiment ofV the present invention is shown in the YformY `of two neutronic reactors utilizing as Vthe'reactiye composition thereinaqueous (H2O) solutions of uranyl sulphate.

(UO2SO4) containing about 14.6 percent of isotope uranium 235 instead of 0.7% as in natural uranium.V

In the drawings l c jFig. 1V is a vertical view partly in section and partly in elevation of a neutronic reactor which has been constructed and is adapted to operate at a one watt output illustrating the present invention; Y j

Fig. 2 is a cross-sectionalrview taken as indicated' by the line 2-2 in Fig. l;

trol rod R becomes 1.001, and ,assumer that the system exponential rise in neutron density, then l Y, Y .l Y, 0.001 1 1 continues indefinitely. The above formula thus indi- I Vthat is, N/N0v-2.72 in 28.5 seconds.l vHence;doublingfrofv Y theV neutron density occurs about everyf20 seconds andl *L Fig. 3 is an enlarged longitudinal sectional view of the Y control rod shown in Fig. 1; Y Figs.4, 5 and 6 are cross-sectional views, as indicated, of'tliecontrolrod; Y Y j }Fig' 7 iis'v a Adiagram showing the solution handling' sys'- tem;"v d n 8 `is Ia diagram showing the electrical control and monitoring system; n y f Fi'gf'9-isafragmentary diagrammatic side Viewv partly Vinisec'tionv ofthe temperature control system and reactor room;

Figiv v1() is n partly in section of the reactor room;

Fig-.'11 is'afchart or graph showing the relation between;

the added amount of uranium 235 and thedepth of insertion of the control rod;

Fig. 12 is a chart or graph showing the effect of ,added uranium 23.5 Aonthe reciprocal period; Y

Fig. 13 is a vertical sectional view of'aulraniurnrZS-SV solutionreactor capable of operating continuously at 'substantial power forexaniple 101m.;V Y f a diagrammatic View partly in topY planr andv L Fig. 14 is a vertical sectional view of the device of Fig. 13 taken in a plane at a right angle to the plane of the section in Fig. 13;

Figure 15 is a graph showing the relation between the critical mass in grams of U235 as a function of the radius of a sphere in accordance with Tables I and III; and,

Figure 16 is a graph showing the concentration in atoms of U235 per molecule of water as a function of the radius of a sphere in accordance with Tables I and III.

Referring tirst to Figs. l and 2, a reactor tank of spherical form is provided approximately 12 inches in diameter and having a volume of 14.95 liters, made of type 347 18-8 stainless steel, Which is sufficiently thin, for example 1,452 inch thick, to absorb but minor amounts of neutrons. The sphere is made from two spun hemispheres with a 'D716 inch equatorial flare, and the hemisphere flares are welded together. Polar flares are also provided, to one of which is welded a top pipe 11. A bottom pipe 12 is welded to the other flare. The top pipe is 11/2 inches inside diameter with a 1,46 inch wall and the bottom pipe is 3%: inch outside diameter with a /s inch wall. Unless otherwise specified hereinafter, all piping in the solution system is of stainless steel.

Referring first to bottom pipe 12. This pipe extends downwardly through a heavy frame base 14 and then through the top of an inverted conical pan 15 to terminate inside thereof just above the bottom point of the pan. Pan 15 is supported on risers 16, which also partially support base 14. Pan 15 can be emptied by a dump pipe 17 under the control of a dump Valve 19 having an extension handle 20. A funnel 21 is provided through which contents of sphere 10 and pan 15 can be conducted into a sump 22, when dump valve 19 is open. In view of the neutronic reactivity of the solution to be used in reactor tank 10, tank 22 may be provided with neutron absorbers such as cadmium baies 24, to prevent neutronic reaction therein.

Top pipe 11 extends upwardly through a cross-frame member 25, this cross-frame member being supported by uprights 26 resting on frame base 14. Above cross-frame member 25 upper pipe 11 terminates in an expanded portion 27 provided with a removable cap 29. An overiiow pipe 30 is provided leading outwardly from expanded portion 27. The remainder of the liquid handling system will be explained later.

Inasmnch as a very considerable weight will be placed on base 14, base 14 is additionally supported by base uprights 32. A reflector base 34 formed from carefully machined graphite bricks is piled on base 14, this graphite being of high neutronic purity. Resting on graphite base 34 and surrounding reactor tank 10 is a reflector 35 of beryllium oxide bricks having a density of about 2.7 gms./cm.3, carefully finished to iit together with a minimum of air spaces of maximum neutronic purity, and with bricks adjacent the reactor tank 10 being shaped to the contour of the tank. The beryllium oxide reector is roughly of spherical shape to provide a neutron reecting layer around the reactor tank. Before assembling the reliector around the reactor tank, means for detecting leaks in the tank are provided in the form of small, preferably nylon insulated, copper wires 36 Wound around the tank 20. While only a single circuit is shown, separate circuits can be used for the top, equator and the bottom of the reactor tank, if desired. If a leak from the tank occurs, the solution will saturate the insulation on the wire and ground it to the reactor tank 10, as will be later described. Thermocouples may also be inserted in various positions around the reactor tank, as indicated by thermocouple 37 positioned adjacent the top of the reactor tank 10.

As the reflector 35 is being assembled, two-vertical tangential slots are built into space slightly away from tank 10 in the reflector, a wide control rod slot 40 and a safety rod slot v41 close to tank 10. Both of these slots may be provided with an aluminum lining or Yscabbard attached 12 to the equator of tank 10. Operating in the control rod slot 40 is a control rod 42. The control rod proper is a strip of .032 inch cadmium 34 inches long, wrapped around a hollow brass tube 3A inch in diameter and 34 inches in length, and is moved in a vertical direction with a total length of motion of 40.7 inches by a control rod motor 44, the position of the rod being indicated by Selsyn repeater 45. The details of this control rod mechanism is shown more in detail in Figs. 3 to 6 inclusive, and will next be described.

As shown in Fig. 3, a screw shaft 46 is mounted vertically in a screw shaft bearing 47 mounted on top frame member 48 and extends upwardly to receive a spur gear 49 pressed against a shoulder 50 keyed to shaft 46 by clutch spring 51 retained by end nut 52. Clutch spring 51 forces clutch plate 53 against spur gear 49 and-spur gear 49 against shoulder 50. Spur gear 49 is driven by the reversible D. C. control rod motor 44 through pinion 55. Thus screw shaft 46 is rotated by the motor 44 through a friction clutch drive. The lower portion of shoulder 50 is provided with pinion teeth 56 engaging a driven spur gear 57 attached to the shaft of the Selsyn repeater 45.

Extending downwardly from bearing 47, is a rod casing 59 terminating in a casing block 60, which also supports an upwardly extending inner tube 61. Immediately inside of inner tube 61 is a control rod sheath 62 which extends all the way from bearing 47 to the full desired extent of control rod motion in control rod slot 40. Sheath 62 is sealed at the bottom by a welded cap 64.

The control rod proper, as above described, consists of a cadmium layer 65, sandwiched between inner and outer brass tubes 66 and 67, respectively, these tubes being attached at their upper end to a nut 69 sliding inside of rod sheath 62 and prevented from turning by a projection 70 entering aligned slots 71 in tubes 61 and 62. Nut 69 is .threaded on threads 72 cut on the portion of shaft 46 below bearing 47. Thus rotation of shaft 46 will raise and lower the control rod within a watertight sheath. This watertight construction is not important when a beryllium oxide reector is utilized, but is useful in case a liquid retiector, such as deuterium oxide (heavy water), is used.

The safety rod 76 (Figs. l and 2) consists of a cadmium sheet .O32 inch thick, 21/2 inches wide and 42 inches long, sandwiched for strength between two similar pieces of brass. In its bottom position, its lower end extends 8 inches below the center of the tank 1t). Normally an electromagnet 77 holds the safety rod out of the reiiector by means of a safety rod disc 79 of magnetic material attached directly to the top of the rod. Any interruption of current in the magnet, brought about either manually or by means of any of the safety circuits, later to be described, will release the rod to fall freely into the reflector by gravity. A tripping switch S0 is provided just above the top position of the magnet 77 so that if the magnet should be lifted too high, the safety rod will be dropped. In addition, upper and lower position indicator switches 81 and 82, respectively, are provided so that the in or out position of the safety rod can be made known to the operator of the reactor.

The safety rod is raised and lowered as desired by a safety rod motor 84 operating a drum 85 winding a cable 86 attached to the electromagnet 77. VThe safety rod motor 84 is a standard reversible motor, and is operated through circuit R which is connected to a Switch 148 located in the control room. The switch 148 is a standard switch having three positions, an off position, an electromagnet raising position, and an electromagnet lowering position. Limit switches 89 and 90 are provided, operated by a stop on the drum 85, to limit the top and bottom respectively, of the safety rod travel. Limit switch is an additional safeguard in case limit switch 89 does not operate to stop motor 84. A sliding brake 91 is provided on the safety rod to soften the blow on the sti-ue position of brakeV 91. Y

4Certain other safeguards are attached to the system as so far described, and while their position Willbe indicated I here, their functions will be taken up later. For eXam- Y ture .when the rod is' dropped.z It may consist of two brass bars clamped tothe rod with the friction adjusted by a'spring. Normally, this brake is about 4 inches-from the top of the rod. The rod, therefore, falls freely when rreleased by magnet 77 until the brake 91 hits cross-frame member 25 after whichthe rod has to vslide between the brass bars with some friction. A stop 92 is provided nearthe top of the structure to prevent the brake from rising. all thev way, and for the last 4 inches the safety rod has to slide between the brass bars, thus resetting the ple, immediately below the expanded portion 27 on the top of the upper pipe 11 area pair ofsolution contact switches 95 and 96, switch 95 being slightly lower than switch96. These switches are used to monitor Ithe'upper level of solution in the reactor system. Such switches' ground a central electrode to the system through the solutio'n,'whichY is velectrically conductive. A lower level indicator switch 97 of similar type is provided on lower pipe 12 just above the top of conical pan 15.V The top of pan 15 is provided with a pan air supply line 99 and an electrically operated air release valve 100, and a level indicator 101. Pan 1,5 is also provided with a pan thermocouple 102 for determining the temperature ofthe liquid in the pan 15.

`Neutron monitoring ionization chambers are also provided. A- pair of BFS ionization chambers 105 and 106 (Fig. 2) are provided outside of the reflector 35 and positioned behind a lead shield 107 where the chambers willstill receive a sufcient neutron density during operation of the reactor to give, proper monitoring of the neutron reactivity. An additional ionization Vchamber 109 (Fig. l) is provided adjacent pan 15 to monitor the radiationractivity of the liquidy in this pan. j'

`The liquid handlingsystem as shown in Fig. -7 Will next be referred to.V Inasmuch as one preferred solution to "bensed in the reactor is a uranyl sulphate solution 'in ordinarywater, with the uranium 235 content of the uranium much higher than in naturaluranium, it is important that evaporation from the solution -be controlled tially constant in concentration during Vuse and therefore -is being separated from the operating air. To attain this result, and to lill reactor tank 10, a source of compressedv f aijis provided, arrivingthrough air pipe v110.*under the control of system inlet valve 111 and balloon lling valve 112 (Fig. 7 Between the two valves a supply pipe 114 leads to an air reservoir 115. -an'air pressure gauge -116, and an electrically operated air dump valve 117. In the interior-of air reservoir'llSV are positioned flexible balloons 119 connected topan aii supplypipe 99 (Fig. l) line and to a manometer 121 havferred that valves 111 and 112, `air pressure gauge 116, air dump valve 117 and manometer121 be positioned within a control rocrn, as indicated by enclosure line 125. All air lines are of /{16 O. D. copper. .Valve handle 20 is also extended to this room. Y'

j Itcan be seen'from the air linevcir'cuit so fardes'cribed,I

that air pressure -can be applied to'the top of a solution Y positioned in pan 15 through the 'medium yof theA balloons charged 'with air through valve 112.

""t the top Vof the system, overflow pipe 30 (Fig. 7)

'119,y and that the system and balloons can be originally Attached to pipe 114 isV 147 opening air release valve 117, is also operated by.

4s'that the solution being handled may remain substang thermocouples 37 and 102 operate temperature indicatorsV 152 and 154 respectively, Ain the control room through inga'liquid level indicator switch 122 therein; It is prereaction tank 10 and beyond tank 10 into upper pipe 11. The operation of the .reactor will beA described after the electrical monitor system has been described, as shown in Fig. 8, which vis highly Ydiagrammatic and reduced to lowest terms. AThe right hand side of the diagram shows the instruments in the control Vroom, the left hand `side indicates the circuit. The lettersdenote the continuity Yof the circuits with those shown in other figures.y l l The three BF3 ionization chambers, i. e., chambers 105 and 106, placed adjacent the reactor, see Figure 2, and the chamber 109 placed adjacent Ythe pan 15, see Figure l,

energize D. C. ampliers 135, 136 and 139 respectively.Y

The outputs from these amplifiers operate-respective monitoring galvanometers a, 136a, and 139:1, to indicate radiation values. The output circuit from these amplifiers 135 and 136 also pass through tripping circuit relays as indicated by numeral 140. The tripping irnpulse, carried on trip line 141, operates a relay in a safety rod magnet power supply indicated by numeral 142,A to break current carried inelectromagnet circuitA, connected to electromagnet 77 holding up the safety rod 76. The safety rod magnet power supply 142 can also be interrupted by a hand switch 144, and by the upper safety rod limit switch 80 through circuit B. The tripping impulse from the trippingrcircuit also is passed through a time delay relay (about 2 seconds delay) indiby numeral 147, the output of Vwhich is carried by circuitY C, to the electrically operated air release valve 117, which is in the'control room. The solenoid air' release circuit loverflow contact switch 127 (Fig. 7) through circuit D; by leak detector wire 36 through circuit E (Fig..1); and by upper liquid level switch 96, through circuit F (Figs. l and 7). It will be noted in this respect that the overliow contact switch 127 backs up upper level switch 96 as a safeguard in case switch'96 may Vnot operate properly, as, for example, when the solution is blown overl into overow tank 126 by an air bubble, etc.

The control rod motor 44 is operated by in and out switch through circuit G and Selsyn 45 drives a control rod depth indicatorv 151 in the control :roomVV through circuit H. Y

The various thermocouples such as, Vfor example,

circuits I and J respectively. Further, in order to ensure emergency release of the air pressure in the system, in case valve 117 does not operate properly, the electrically operated valve 100 on top of pan 15 canv be opened by opening manual emergency switch 155. Normally Vthe valve 100 is held closed by power Vfrom themains through circuit K, but Yupon failure of power in the mainsL the valve is kept closed by a separate D. C. source 156.

All other safety circuits are arranged to operate upon power failure in the mains. Y u Y Safety rod positionindicating switches 82 and 81 operate respective in and out lights 157 andV 159 in theY control room through circuits L and M, respectively.` l Pan level switch 101 operates indicator A16 0' in the control room through circuit N. The liquid level indicator switch: 122 in manometer 121 (Fig. 7) operates lampr161 through circuit O, and the high and low solution levelsV Y in the system are indicated by lamp 162, operated by solution switch 95 throughV circuit P forthe high position,

and by'lamp 164 operated kfromsolution switch 97 Y through circuit Q for the Vlow position. 'Solution position indicating switches 95, 96, 97 and 101 are operated v with circuits normally open to prevent electrolysis of the solution and are checked by push buttons.`

Y Having thus describedl the reactor and the 'control and safety systems, the liquid handling system and the elec- A"N5V trical monitoringand operating system, the operationv of the device will be described considering first that the reactive composition to be used in the reactor is a uranyl sulphate solution having a sufcient uranium 235 concentration to cause the system to be chain reacting when it has filled the 12 inch sphere. This uranyl sulphate solution is to be stored in pan 15 and level switch 101 will detect solution level 124, see Figure 7, thereby indicating if there is sufiicient solution in the pan to fill the reactor tank and pipes. To describe the process of filling the reactor tank and initiating the reaction, reference is made to Fig. 7. The control and safety rods are fully inserted. Then the balloons 119 are filled by use of the valve 112 which is then closed. Air pressure is then admitted to air reservoir 115 through valve 111, compressing the air in baloons 119, causing the pressure to be transferred to the top of the solution in the pan 15. The solution then is forced upwardly in pipe 12, into reactor tank 10 and then into upper pipe 11. The progress of the rise of the solution through the system may be checked by watching manometer 121 which is calibrated to give a rough position of the liquid level as it is rising. This manometer can also be used to detect air leaks in the system. The upper solution switch 95, mounted on upper pipe 11, indicates when the solution reaches that level, which is the normal operating level of the solution in the system.

It will be noted that the solution will stay at this level only if electromagnetic air release valves 117 and 10i) `remain closed. If either of these valves are opened, the

air pressure on the solution is released and this release will permit the solution to flow back into conical pan 15, and out of the tank 10, by gravity. It will be noted from Fig. 8 that the solution will be dumped by operation of valve 117 through circuit C, when any one of a number of things happen. First, if the solution level rises beyond upper solution level switch 96, second if there is an overflow into overflow tank 126, and third, ifjthere is a leak in the tank itself, such as would saturate the nylon covering of the wire 35 wound around the reaction tank. Furthermore, valve 117 will also be operated two seconds after the tripping circuit 140 has created an impulse to drop the safety rod. Therdelay in this case is to prevent the solution from being dumped if the entrance of the safety rod into the reliector properly stops a rise in neutron density, as would be indicated by either of the ionization chambers 195 or 1%6. if the safety rod does not stop the rise in neutron density after 2 seconds, the solution is then automatically dumped out of the sphere. Finally, if none of these automatic precautions operate, the manual emergency dump circuit K can be operated to release the air through the solenoid operated air valve This last operation is only used as a final emergency procedure, as fresh air must then be supplied to the entire system. All dumping circuits vexcept the emergency circuit K are arranged so that if power supply fails, solenoid valve 117 will open.

As final precaution, if the solution is too highly radioactive as indicated by ionization chamber 109, and galvanometer 139m the conical pan 15 itself can be dumped by use of the manually operated valve handle 2i), preferably extended to the control room so that the solution can be conducted outside the operating room into storage tank 22. The cadmium baflles 24, being strong neutron absorbers, effectively prevent any possibility of the chain reaction taking place in tank 22.

A discussion of some of the nuclear aspects of the system will be given prior to describing the start-up of the operation of the device.

While the entire volume of the solution Vis normally stored in conical tank 15, no chain reaction will take place therein for several reasons. First, the sphere is the most eflicient shape for a neutronic reactor, Whereas the conical shape is not. Second, neutronic reactors of small size have an extremely high neutron leakage' factor. When a reflector is used around tank 10, critical mass can be obtained with a lower concentration of uranium 235, because the reflector returns neutrons to the solution and very effectively reduces the amount ofuranium 235 required in the tank 10 to cause the chain reaction to be attained. Since conical pan 15 has no reector, most of the escaping neutrons are lost and do not return. In consequence, no chain reaction takes place in pan 15.

However, the solution in pan 15 can become highly radioative after operation of the device as a neutronic reactor, due to the accumulation of radioactive fission products therein. ionization chamber 109 is used to monitor this radioactivity and if it becomes too high, the solution may have to be drained into storage tank 22 until the radioactivity decays to a safe handling value. Alternatively an auxiliary tank may be provided as a substitute for tank 15.

Other features should be pointed out. It will be noted that neither the safety rod nor the control rod enter the reactor tank 1t). Small reactors such as shown and described herein have sueh high neutron leakage that they usually are not of critical size without a reilector and are dependent for proper operation for a given size, concentration and shape, on the eflicient action of the reflector. In such a small reactor the insertion of neutron absorbers even in the reflector outside of the reactor tank will prevent the reflector from returning sufficient neutrons to keep the chain reaction sustained with the reactor having a mass that would be critical if it were not for the absorption in the reflector. This affords a very simple and effective method of control without insertion of neutron absorbers into the reacting portion of the reactor.

Of the uranium salts, UO2SO4 is preferred for use in the reactor instead of, for example, uranyl nitrate, first, because there is less unwanted neutron absorption with the suphate than there is with the nitrate, and, second, the sulphate is more soluble in water than the nitrate. Furthermore, 18 -8 stainless steel-has showed extremely low corrosion rates after being in contact with UOZSO., solutions from l to 2 weeks. Consequently, all portions of the system which are to come into contact with UO2SO4 are pickled with normal 3 M UO2SO4 solution for from l to 3 weeks before starting operations.

In starting up the device for the first time, a suicient amount of distilled water is placed in conical pan 15 to properly ill the reactor tank and its attached pipes to the proper operating level as indicated by solution switch 95. Uranyl sulphate containing isotope uranium 235 to the point where the average computation of the material is about 14.7 percent uranium 235, as indicated by mass spectrometer analysis, is added to l or 2 liters of the water withdrawn from the conical reservoir, and dissolved therein. The resulting solution is replaced in the tank 15 and stirred as, for example, with an electric mixer, through a cap Vto prevent evaporation of the water. The solution is then run up and down in the reactor tank 10 several times, while the control rods are in, to irnprove the .mixing When the neutron counting rate, as indicated by monitoring counters and `106 does not change with each successive filling of the tank 10, the solution is adequately mixed. This method of adding the salt increases the total volume of the solution at each step, and to avoid accumulation of too much excess solution, some of the solution is removed during the addition, evaporated, and the recovered UO2SO4 made ready forfurther use.

To establish a chain reaction uranyl sulphate isadded in the manner described, until critical conditions are reached, i. e., where the neutron reproduction ratio in the reactor tank equals unity. With the l2 inch reactor tank, critical conditions are obtained with about 570 grams of uranium 235.

With the cadmium control Vrod partly in, some uranium 235, such as approximately 1.8 grams, is removed from 

1. A NEUTRONIC REACTOR COMPRISING A VESSEL HAVING ALL DIMENSIONS FROM ITS CENTER TO ITS WALL OF ABOUT 6 INCHES, A LIQUID, SAID VESSEL SUBSTANTIALLY FILLED WITH SAID LIQUID, SAID LIQUID COMPRISING AN ORDINARY WATER SOLUTION OF A SOLUBLE SALT OF URANIUM, SAID URANIUM HAVING AT LEAST ABOUT ONE PERCENT BY WEIGHT OF THE FISSIONABLE ISOTOPE U235, SAID SOLUTION HAVING A CONCENTRATION OF GREATER THAN ABOUT 10-3 ATOMS OF SAID FISSIONABLE ISOTOPE PER MOLECULE OF WATER THE QUANTITY OF FISSIONABLE ISOTOPE IN SOLUTION BEING IN EXCESS OF ABOUT 500 GRAMS AND SUFFICIENT TO PRO- 